ML20078G698
| ML20078G698 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 10/03/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20078G691 | List: |
| References | |
| NUDOCS 8310130079 | |
| Download: ML20078G698 (5) | |
Text
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NUCLEAR REGULATORY COMMISSION j
W ASHINGTON, D. C. 20555 5
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR P.EACTOR REGULATION SUPPORTING AMENDMENT N0S. 66 AND 60 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT,. UNIT NOS. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 Introduction By letter dated June 24, 1983 (Ref. 1), Northern States Power Company (NSP) requested changes to the Technical Specifications (TS) for the Prairie Island Nuclear Generating Plant Unfts Nos, and 2.
The proposed TS changes involve the local power hot channel factor F limit in t e TS 3.10-1, 3.10-2, 3.10-9 and 3.10-11 where (1) a numerical va ue in the F limit expression will be changed from 2.21 to 2.32; and (2) the normalize exposure dependence function Bu(z) curve will be changed to 1.0 for a 1 values of peak pellet exposure from 0_to 55 GWD/MTU. The change involvi g F applies only for fuel manufactured by the Exxon Nuclear Company. The F fa tor would remain the same for the stinghouse manufactured fuel (i.e.
2.21).
In addition, the definition of F will be changed from a " neutron flux comparison" to " heat flux comparison".
I its letter of July 29, 1983 (Ref. 2), NSP also proposed a change to the axial dependence function K(z) in Figure TS 3.10-5.
These TS change requests are based on a new LOCA analysis performed by the Exxon Nuclear Company for Prairie Island. The analysis results are described in the report XN-NF-83-38 (Ref. 3). The staff evaluation regarding the TS changes follows.
Staff Evaluation The purpose of limiting condition of operation on Fk is to limit the peak linear power density during normal operation to provide assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS accept-ance criteria limit of 2200 F is not exceeded.
In order to support the proposed TS changes, the licensee in the XN-XF-83-38 report (Ref. 3) describes the analysis and results performed by Exxon for the Prairie Island Units 1 and 2 for a postulated large break LOCA. The analysis was made with the following conditions:
(1) The double-ended cold leg guillotine break with a discharge coefficient of 0.4: This scenario has been identified in the previous analyses as the most limiting break.
(2) An entire core with ENC TOPROD fuel: With respect to LOCA, the TOPROD fuel design is more limiting than the ENC XN-1 and XN-2 fuels due to the smaller pin diameter and the increased core flow area which reduces core reflood rates in the LOCA analysis.
0310130079 831003 PDR ADOCK 05000282 p
.. (3) 5%,f steam generator tubes being uniformly plugged.
(4) Including 2% power uncertainty resulting in the allowable linear heat generation rate of 15.02 kW/ft corresponding to total power peaking factor of 2.32 and nuclear enthalpy rise factor of 1.55 for the entire fuel exposure.
(5) Maximum peak pellet exposure of 55 GWD/MTU.
The reactor coolant system nodalization was modeled in accordance with an approved ENC ECCS modeling described in XN-NF-77-25(A) (Ref. 4) for a 2-loop Westinghouse PWR with dry containment. The LOCA analysis was performed using ENC's EXEM/PWR ECCS evaluation model (Ref. 5). This evaluation model uses the following computer codes:
(1) RELAP-EM (Ref. 6) for the system blowdown and hot channel blowdown calculations; (2) CONTEMPT-LT/22 as modified in CSB G-1 (Ref. 7) for the containment back pressure calculation; (3) REFLEX (Ref. 8) for system reflood calculation; (4) RODEX2 (Ref. 9) for initial fuel rod stored energy, fission gas release and internal gas inventory calculations; and (5) T00DEE2 (Ref.10) for the calculation of final fuel rod heatup.
The RELAP-EM, CONTEMPT-LT/22, REFLEX and T00DEE2 codes have previously been approved by the NRC. The R0DEX2 code has recently been reviewed by the staff and has been found acceptable for use in the LOCA initial stored energy and rod pressure calculations (fonnal SER for R0DEX2 is being prepared).
In addition, an approved cladding swelling and rupture model described in XN-NF-82-07, Revision 1 (Ref. 11) was used in the calculation of the cladding rupture, strain and flow blockage in the ENC's EXEM/PWR ECCS evaluation model and, therefore this part of the analysis is acceptable. The overall EXEM/PWR ECCS evaluation model is still under review by the staff. However, the review has progressed to the point to conclude that the evaluation model is acceptable for the Prairie Island LOCA analyses.
The LOCA analysis was performed with two analyses:
from the beginning of life to 15 GWD/MTV and from 15 GWD/MTU to 55 GWD/MTV. The most limiting fuel conditions in the respective exposure ranges were used in the analysis.
The combination of the highest stored energy, rod pressure and decay power was used to bound the LOCA-ECCS analysis over the exposure ranges. The results of these analyses have shown the maximum peak cladding tempera-tures to be 2091*F and 2142'F, respectively, for the fuel exposure ranges prior to and after 15 GWD/MTU. The local metal-water reactions are 4.68%
and 5.6% for the two fuel exposure ranges, and total core metal-water
.. reaction is less than 1%. These peak clad temperatures (PCT) and metal-water
)
reaction are within the limit imposed in 10 CFR 50.46, " Acceptance criteria for ECCS for 1 ght water nuclear power reactors". Therefore, the proposed l
TS change in F limit is acceptable.
The staff also reviewed the change to Figure TS 3.10-5 proposed in the July 29, 1983 letter. This curve is a core axial dependent, function, usually referred to as the K(a) curve. The K(a) curve should be revised whenever a large or small break LOCA analysis is performed with a changed peaking factor. We have independently determined that the curve submitted in the July 29, 1983 letter (not the curve submitted in the June 24, 1983 letter) is correct.
This proposed TS change is therefore acceptable.
WithregardtotheproposedchangeinthedefinitionofFNfroma" neutron flux comparison" to a " heat flux comparison", we find thaY the heat flux comparison is more appropriate since the concern of PCT in the LOCA analysis is directly related to the thermal power. Therefore the proposed change is acceptable.
In coDelusion, the staff has reviewed the NSP proposed TS changes regarding theFflimitingconditionofoperationandfoundthemacceptableforonlythe Exxon manfactured fuel. This acceptance is based on the LOCA analysis results provided by the licensee using the approved computer codes, and the staff's review of the ENC's EXEM/PWR ECCS evaluation model used in this analysis has progressed to the point where the evaluation model is acceptable for the Prairie Island LOCA analysis. Therefore, although a reanalysis could be re-quired if the final evaluation of the EXEM/PWR ECCS model by the staff identi-fied any problems or condition, we have concluded that the proposed change in the Prairie Island Units 1 and 2 Technical Specification is acceptable.
Since the effects of the increase in the fuel exposure to 55 GWD/MTU on the l
off-site doses have not been addressed here, the staff's acceptance does not include the limitation of fuel burnup to 55 GWD/MTU.
Since the proposed TS change is based on the LOCA analysis results showing that the accaptance criteria imposed in 10 CFR 50.46 will not be violated, the staff has determined that these TS changes will not involve a significant reduction in safety margin.
Environmental Consideration l
We have determined that the amendments do not authorize a change in effluent l
types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is l
insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement or negative declara-tion and environmental impact appraisal need not be prepared in connection with the issuance of the amendments.
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Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: October 3, 1983 Principal Contributors:
Yihsi,ung Hsii Marvin Dunenfeld Shih-Liang Wu 4
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.. References 1.
Letter from D. Musolf (NSP) to NRC, " Prairie Island Nuclear Generating Plant, Docket Nos. 50-282, 50-306, License Amendment Request dated June 24, 1983, LOCA Analysis", June 24, 1983.
2.
Letter from D. Musolf (NSP) to NRC, " Revision No. I to License Amendment Request dated June 24, 1983 LOCA Analysis", Docket Nos. 50-282, 50-306, July 29, 1983.
3.
XN-NF-83-38, " Prairie Island Units 1 and 2 Limiting Break LOCA-ECCS
' Analysis Using EXEM/PWR", Exxon Nuclear Company, May 1983. Exhibit C To NSP Junc 24, 1983 letter.
4.
XN-NF-77-25(A), " ENC ECCS Evaluation of a 2-loop Westinghouse PWR with Dry Containment Using the ENC WREM-II ECCS Model - Large Break Example Problem", September 1978.
5.
XN-NF-82-20, " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates", Revision 1, August 1982; Supplement 1, March 1982; Supplement 2, March 1982.
6.
Letter from T. A. Ippolito (NRC) to W. S. Nechodam (ENC), "SER for ENC RELAP4-EM Update", March 1979.
7.
NRC Branch Technical Position CSP G-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation".
8.
Memorandum from R. L. Tedesco to B. K. Grimes, " Review of Exxon Nuclear Company Topical Report (ENC WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA), TAC-10691", March 23, 1979.
9.
XN-NF-81-58(P), Revision 2:
Fuel Rod Thermal-Mechanical Response Evaluation Model", February 1983.
- 10. NUREG-75/057, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program", May 1975.
- 11. XN-NF-82-07(P), Revision 1, " Exxon Nuclear Company ECCS Claddin9 Swelling and Rupture Model", August 1982.
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